Federal Register / Vol. 63, No. 17 /Tuesday, January 27, 1998 / Notices 3927

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. STN 50–454, STN 50–455 and

STN 50–456, STN 50–457]

Commonwealth Edison Company;

(Byron Station, Units 1 and 2);

(Braidwood Station, Units 1 and 2);

Exemption

I

Commonwealth Edison Company

(ComEd, the licensee) is the holder of

Facility Operating License Nos. NPF–37,

NPF–66, NPF–72, and NPF–77, which

authorize operation of Byron Station,

Units 1 and 2, and Braidwood Station,

Units 1 and 2, respectively. The licenses

provide, among other things, that the

licensee is subject to all rules,

regulations, and orders of the

Commission now or hereafter in effect.

The Byron facility consists of two

pressurized-water reactors located at the

licensee’s site in Ogle County, Illinois.

The Braidwood facility consists of two

pressurized-water reactors located at the

licensee’s site in Will County, Illinois.

II

In its letter dated April 3, 1997, as

supplemented on June 19, 1997, ComEd

requested an exemption from the

Commission’s regulations. Title 10 of

the Code of Federal Regulations, Part 50,

Section 60 (10 CFR 50.60), ‘‘Acceptance

Criteria for Fracture Prevention

Measures for Lightwater Nuclear Power

Reactors for Normal Operation,’’ states

that all lightwater nuclear power

reactors must meet the fracture

toughness and material surveillance

program requirements for the reactor

coolant pressure boundary as stated in

Appendices G and H to 10 CFR Part 50.

Appendix G to 10 CFR Part 50 defines

pressure-temperature (P-T) limits during

any condition of normal operation,

including anticipated operational

occurrences and system hydrostatic

tests to which the pressure boundary

may be subjected over its service

lifetime, and specifies that these P-T

limits must be at least as conservative as

the limits obtained by conforming to the

methods of analysis and the margins of

safety of the American Society of

Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code (Code),

Section XI, Appendix G. 10 CFR 50.55a

requires that any reference to ASME

Code Section XI in 10 CFR Part 50 refers

to addenda through the 1988 Addenda

and editions through the 1989 Edition of

the Code unless otherwise noted. It is

specified in 10 CFR 50.60(b) that

alternatives to the requirements

described in Appendices G and H to 10

CFR Part 50 may be used when an

exemption is granted by the

Commission under 10 CFR 50.12.

To mitigate low-temperature

overpressure transients that would

produce pressure excursions exceeding

the required limits while the reactor is

operating at low temperatures, the

licensee installed a low-temperature

overpressure protection (LTOP) system.

The system includes pressure-relieving

devices called power-operated relief

valves (PORVs). The PORVs are set at a

pressure low enough so that if an LTOP

transient occurred, the mitigation

system would prevent the pressure in

the reactor vessel from exceeding the

required limits. To prevent the PORVs

from lifting as a result of normal

operating pressure surges, some margin

is needed between the PORV setpoint

and the normal operating pressure. In

addition, when instrument uncertainty

is considered, the operating window

between the PORV setpoint and the

minimum pressure required for reactor

coolant pump seals is small and

presents difficulties for plant operation.

The licensee has requested the use of

the 1996 Addenda to the ASME Code,

Section XI, Appendix G, which allows

the use of lower stress intensity factors

for determining the applied stress

intensity from pressure and thermal

stresses, and allows use of an LTOP

system setpoint so that system pressure

does not exceed 110 percent of the P-T

limits. The 1996 Addenda to the ASME

Code, Section XI, Appendix G, is

consistent with guidelines developed by

the ASME Working Group on Operating

Plant Criteria to define pressure limits

during LTOP events that avoid certain

unnecessary operational restrictions,

provide adequate margins against failure

of the reactor pressure vessel, and

reduce the potential for unnecessary

activation of pressure-relieving devices

used for LTOP. ASME Code, Section XI,

Appendix G, 1996 Addenda, has been

approved by the ASME Code

Committee.

III

Pursuant to 10 CFR 50.12, the

Commission may, upon application by

any interested entity or upon its own

initiative, grant exemptions from the

requirements of 10 CFR Part 50 when (1)

the exemptions are authorized by law,

will not present an undue risk to public

health and safety, and are consistent

with the common defense and security;

and (2) when special circumstances are

present. Special circumstances are

present whenever, according to 10 CFR

50.12(a)(2)(ii), ‘‘Application of the

regulation in the particular

circumstances would not serve the

underlying purpose of the rule or is not

necessary to achieve the underlying

purpose of the rule. * * *’’

The underlying purpose of 10 CFR

50.60 and 10 CFR Part 50 Appendix G

is to establish fracture toughness

requirements for ferritic materials of

pressure-retaining components of the

reactor coolant pressure boundary to

provide adequate margins of safety

during any condition of normal

operation, including anticipated

operational occurrences, to which the

pressure boundary may be subjected

over its service lifetime. Section IV.A.2

of Appendix G to 10 CFR Part 50,

requires that the reactor vessel be

operated with P–T limits at least as

conservative as those obtained by

following the methods of analysis and

the required margins of safety of

Appendix G of Section XI of the ASME

Code. 10 CFR 50.55a requires that any

reference to ASME Code Section XI in

10 CFR Part 50, Appendix G, refers to

addenda through the 1988 Addenda and

editions through the 1989 Edition of the

ASME Code, unless otherwise noted.

Appendix G of the ASME Code

requires that the P-T limits be

calculated: (a) Using a safety factor of

two on the principal membrane

(pressure) stresses, (b) assuming a flaw

at the surface with a depth of onequarter of the vessel wall thickness (1⁄4

T) and a length of six (6) times its depth,

and (c) using a conservative fracture

toughness curve that is based on the

lower bound of static, dynamic, and

crack arrest fracture toughness tests on

material similar to the reactor vessel

material.

For determining the P–T limits, the

licensee proposed to use the safety

margins based on the 1996 Addenda to

the ASME Code in lieu of the 1989

Edition. When compared to the 1989

Edition of the ASME Code, the 1996

Addenda permits the use of a lower

stress intensity factor for determining

the applied stress intensity from

pressure and thermal stresses. This

results in a slight reduction in the

applied stress intensity and a

corresponding shift in the allowable

pressure at a given temperature in the

non-conservative direction; however,

this difference is minor when compared

to the explicit conservatisms

incorporated into Appendix G, and the

changes in the stress intensity factor are

supported by the work performed for

NRC and for others by J.A. Keeney and

T.L. Dickson at Oak Ridge National

Laboratory (ORNL).

For determining the LTOP system

setpoint, the licensee proposed to use

safety margins based on the 1996

Addenda to the ASME Code. The 1996
3928 Federal Register / Vol. 63, No. 17 /Tuesday, January 27, 1998 / Notices

Addenda allows determination of the

setpoint for mitigating LTOP events so

that the maximum pressure in the vessel

would not exceed 110 percent of the

P–T limits that are determined using the

1996 methodology. This results in a

safety factor of 1.8 on the principal

membrane stresses. All other factors,

including assumed flaw size and

fracture toughness, remain the same.

Although this methodology would

reduce the safety factor on the principal

membrane stresses, the proposed

criteria will provide adequate margins

of safety for the reactor vessel during

LTOP transients and, thus, will satisfy

the underlying purpose of 10 CFR 50.60

for fracture toughness requirements.

Further, by relieving the operational

restrictions, the potential for

undesirable lifting of the PORV would

be reduced, thereby improving plant

safety.

It should be noted that the provision

to set the PORV setpoint so that system

pressure remains below 110 percent of

the P–T limits has already been

incorporated into the Byron and

Braidwood licensing basis. This

provision was approved by an

exemption to 10 CFR 50.60 granted to

Byron, Units 1 and 2, on November 29,

1996, to Braidwood, Unit 1 on July 13,

1995, and to Braidwood, Unit 2 on

December 12, 1997, to allow the use of

ASME Code Case N–514. Therefore,

although it represents a change from the

1989 Edition of the ASME Code, it is not

a change to the current licensing basis

for the facilities.

IV

For the foregoing reasons, the NRC

staff has concluded that ComEd’s

proposed use of the alternate

methodology in determining the

acceptable setpoint for LTOP events will

not present an undue risk to public

health and safety and is consistent with

the common defense and security. The

NRC staff has determined that there are

special circumstances present, as

specified in 10 CFR 50.12(a)(2), in that

10 CFR 50.60 need not be applied in

order to achieve the underlying purpose

of this regulation, which is to provide

adequate fracture toughness of the

reactor pressure boundary.

Accordingly, the Commission has

determined that, pursuant to 10 CFR

50.12(a), an exemption is authorized by

law, will not endanger life or property

or common defense and security, and is,

otherwise, in the public interest.

Therefore, the Commission hereby

grants an exemption from the

requirements of 10 CFR 50.60 so that the

P–T limits may be determined using the

1996 Addenda to the ASME Code,

Section XI, Appendix G, and the LTOP

system setpoint may be determined so

that system pressure does not exceed

110 percent of the P–T limits.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this exemption will not have

a significant effect on the quality of the

human environment (63 FR 2268).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 16th day

of January, 1998.

For the Nuclear Regulatory Commission.

Frank J. Miraglia,

Acting Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–1902 Filed 1–26–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–271]

Vermont Yankee Nuclear Power

Corporation, Vermont Yankee Nuclear

Power Station; Exemption

I

The Vermont Yankee Nuclear Power

Corporation (the licensee) is the holder

of Facility Operating License No. DPR–

28, which authorizes operation of the

Vermont Yankee Nuclear Power Station.

The license provides, among other

things, that the licensee is subject to all

rules, regulations, and orders of the

Nuclear Regulatory Commission (the

Commission) now or hereafter in effect.

The facility consists of a single-unit

boiling-water reactor located at the

licensee’s site in Windham County,

Vermont.

II

Section 70.24 of Title 10 of the Code

of Federal Regulations (10 CFR 70.24),

‘‘Criticality Accident Requirements,’’

requires that each licensee authorized to

possess special nuclear material (SNM)

shall maintain a criticality accident

monitoring system in each area where

such material is handled, used, or

stored. Subsections (a)(1) and (a)(2) of

10 CFR 70.24 specify detection and

sensitivity requirements that these

monitors must meet. Subsection (a)(1)

also specifies that all areas subject to

criticality accident monitoring must be

covered by two detectors. Subsection

(a)(3) of 10 CFR 70.24 requires licensees

to maintain emergency procedures for

each area in which this licensed SNM

is handled, used, or stored and also

requires that (1) the procedures ensure

that all personnel withdraw to an area

of safety upon the sounding of a

criticality accident monitor alarm, (2)

the procedures must include drills to

familiarize personnel with the

evacuation plan, and (3) the procedures

designate responsible individuals for

determining the cause of the alarm and

placement of radiation survey

instruments in accessible locations for

use in such an emergency. Subsection

(b)(1) of 10 CFR 70.24 requires licensees

to have a means for identifying quickly

personnel who have received a dose of

10 rads or more. Subsection (b)(2) of 10

CFR 70.24 requires licensees to

maintain personnel decontamination

facilities, to maintain arrangements for

the services of a physician and other

medical personnel qualified to handle

radiation emergencies, and to maintain

arrangements for the transportation of

contaminated individuals to treatment

facilities outside the site boundary.

Paragraph (c) of 10 CFR 70.24 exempts

Part 50 licensees from the requirements

of paragraph (b) of 10 CFR 70.24 for

SNM used or to be used in the reactor.

Paragraph (d) of 10 CFR 70.24 states that

any licensee who believes that there is

good cause why he or she should be

granted an exemption from all or part of

10 CFR 70.24 may apply to the

Commission for such an exemption and

shall specify the reasons for the relief

requested.

III

The SNM that could be assembled

into a critical mass at Vermont Yankee

is in the form of nuclear fuel; the

quantity of SNM other than fuel that is

stored on site in any given location is

small enough to preclude achieving a

critical mass. The Commission’s

technical staff has evaluated the

possibility of an inadvertent criticality

of the nuclear fuel at Vermont Yankee

and has determined that it is extremely

unlikely for such an accident to occur

if the licensee meets the following seven

criteria:

1. Only three new fuel assemblies are

allowed out of a shipping cask or

storage rack at one time.

2. The k-effective does not exceed

0.95, at a 95% probability, 95%

confidence level, in the event that the

fresh fuel storage racks are filled with

fuel of the maximum permissible U–235

enrichment and flooded with pure

water.

3. If optimum moderation occurs at

low moderator density, then the keffective does not exceed 0.98, at a 95%

probability, 95% confidence level, in

the event that the fresh fuel storage

racks are filled with fuel of the

maximum permissible U–235

enrichment and flooded with a